• Nuclear Engineering and Technology
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• 제53권2호
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• pp.549-561
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• 2021

Indonesia has policies and strategies for the management of radioactive waste and spent nuclear fuel that arises from the use of nuclear research and development facilities, including three research reactors, and the use of radioisotopes in medicine and industries. The Indonesian government has provided extensive facilities such as an independent regulatory organization (BAPETEN) and a centralized radioactive waste management organization (CRWT-BATAN). Further, the presence of regulations and several international conventions guarantee the protection of the public from all risks due to handling radioactive waste and spent nuclear fuel. However, the sustainability of radioactive waste management in the future faces various challenges, such as disposal issues related to not only to site selection but also financing of radioactive waste management. Likewise, the problem of transportation persists; as an archipelago country, Indonesia still struggles to manage the infrastructure required for the transport of radioactive materials. The waste from the production of the radioisotopes, especially from the production of 99Mo, requires special attention because BATAN has never handled it. Indonesia should also resolve the management of NORM from various activities. In Indonesia, the definition of radioactive waste does not include NORM. Therefore, the management of this waste needs revision and improvement on the regulations, infrastructure, and technology.

• 방사성폐기물학회지
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• 제19권1호
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• pp.61-74
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• 2021

A paradigm shift in the government's energy policy was reflected in its declaration of early closure of old nuclear plants as well as cancellation of plans for the construction of new plants. To this end, unit 1 of Kori Nuclear Power Plant was permanently shut down and is set for decommission. Based on these changes, the off-site transport of spent fuels from nuclear power plants has become a critical issue. The purpose of this study is to develop an optimized method for transportation of spent fuels from Kori Nuclear Power Plant's units 1, 2, 3, and 4 to an assumed interim storage facility by simulating the scenarios using the Flexsim software, which is widely used in logistics and manufacturing applications. The results of the simulation suggest that the optimized transport methods may contribute to the development of delivery schedule of spent fuels in the near future. Furthermore, these methods can be applied to decommissioning plan of nuclear power plants.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.45-46
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• 2018

• 방사선산업학회지
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• 제17권3호
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• pp.309-320
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• 2023

A multi-purpose canister (MPC) was developed for the purpose of transportation, storage and disposal of spent nuclear fuel (SNF) and has the advantage of minimizing repackaging between management stages of SNF. Considering the typical rock characteristics in Korea, a disposal canister is expected to contain 4 assemblies of Pressurized water reactor (PWR) SNF. The capacity of the MPC should be similarly designed with the disposal canister. However, the MPC with four SNF assemblies is expected to be less efficient in transporting and storing compared to a large-capacity canister. Therefore, a preliminary concept was derived for an auxiliary equipment that can transport and store multiple MPCs in a large overpack. A previously derived concept from US was thoroughly reviewed, and the preliminary concept was revised considering domestic situations including crane capacity and others. In addition, the safety of the normal transportation and storage of the MPC placed in transportation and storage overpack was evaluated with the auxiliary equipment.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.137-138
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• 2018

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.582-589
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• 2010

As a part of the back-end fuel cycle, transportation of spent nuclear fuel (SNF) from nuclear power plants (NPPs) to a fuel storage facility is very important in establishing a nuclear fuel cycle. In Korea, the accumulated amount of SNF in the NPP pools is troublesome since the temporary storage facilities at these NPP pools are expected to be full of SNF within ten years. Therefore, Korea cannot help but plan for the construction of an interim storage facility to solve this problem in the near future. Especially, a decision on several factors, such as where the interim storage facility should be located, how many casks a transport ship can carry at a time and how many casks are initially required, affect the configuration of the transportation system. In order to analyze the various possible candidate scenarios, we assumed four cases for the interim storage facility location, three cases for the load capacity that a transport ship can carry and two cases for the total amount of casks used for transportation. First, this study considered the currently accumulated amount of SNF in Korea, and the amount of SNF generated from NPPs until all NPPs are shut down. Then, how much SNF per year must be transported from the NPPs to an interim storage facility was calculated during an assumed transportation period. Second, 24 candidate transportation scenarios were constructed by a combination of the decision factors. To construct viable yearly transportation schedules for the selected 24 scenarios, we created a spreadsheet program named TranScenario, which was developed by using MS EXCEL. TranScenario can help schedulers input shipping routes and allocate transportation casks. Also, TranScenario provides information on the cask distribution in the NPPs and in the interim storage facility automatically, by displaying it in real time according to the shipping routes, cask types and cask numbers that the user generates. Once a yearly transportation schedule is established, TranScenario provides some statistical information, such as the voyage time, the availability of the interim storage facility, the number of transported casks sent from the NPPs, and the number of transported casks received at the interim storage facility. By using this information, users can verify and validate a yearly transportation schedule. In this way, the 24 candidate scenarios could be constructed easily. Finally, these 24 scenarios were compared in terms of their operation cost.

• 방사성폐기물학회지
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• 제11권4호
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• pp.333-349
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• 2013

최근 국내에서도 원전 부지 내에 건설된 습식저장조의 용량이 곧 포화될 것으로 예상되어 사용후핵연료의 건식저장에 관한 논의가 활발하다. 이 논문에서는 앞으로 다양하게 논의될 저장시스템의 안전성과 함께 장기 건식저장 시 발생하는 사용후핵연료의 특성 및 건전성 변화에 대해 이제까지 국내외에서 연구 보고된 내용들을 면밀히 검토하고 향후 추구해야 할 연구방향을 제시하고자 하였다. 조사 결과 건식저장 기간 동안 진행될 수 있는 여러 피복관 열화기구 중에서 가장 대표적인 기구는 크립 변형과 수소화물에 의한 영향이었으며, 이들이 사용후핵연료 장기 건식저장 시 규제기술기준의 주요 근간을 이루고 있는 것으로 분석되었다. 한편 과거에는 피복관의 크립 변형이 가장 중요한 열화기구로 평가되었으나, 최근의 연구 결과를 통해 수소화물에 의한 영향이 더 심각한 것으로 드러났고 이는 미국의 규제기준과 새로운 온도 범위를 제시하고 있는 일본의 규제기준에서 확인할 수 있었다. 그러나, 아직까지 수소화물에 의한 영향이 발생하는 응력과 온도 조건을 명확히 규명할 수 있는 연구 자료가 충분하지 못하며, 나아가 사용후핵연료의 취급 시 거동에 대한 연구도 지속적으로 수행해야 할 부분으로 드러났다. 따라서 국내 사용후핵연료 특성에 맞는 건식저장조건을 수립하기 위해서는 국내에서도 본격적인 연구를 통해 이들 자료에 대한 충분한 생산과 평가 및 분석이 뒤따라야 할 것으로 판단된다.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.1975-1988
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• 2024

It is important to maintain cladding integrity in spent nuclear fuel management. This study proposes a numerical analysis method to evaluate the fracture resistance of irradiated zirconium alloy cladding under pinch load known to cause Mode-III failure. The mechanical behavior and fracture of the cladding under pinch loading can be evaluated by a Ring Compression Test (RCT). To simulate the fracture of hydride precipitates, zirconium matrix, and Zr/hydride interfaces under the stress field generated by RCT, a micro-structure crack propagation simulation method based on Continuum Damage Mechanics (CDM) has been proposed. Our RCT simulation model was constructed from microscopic images of irradiated cladding. In this study, we developed an automated process to generate a pixel-based finite element model by separating the hydride precipitates, zirconium matrix, and interfaces using an image segmentation method. The appropriate element size was selected to ensure the efficiency and accuracy of a crack propagation simulation. The load-displacement curves and strain energies from RCT were compared and analyzed with the simulation results of different element sizes. The finalized RCT simulation model can be used to establish the failure criterion of fuel rods under pinch loading. The advantages and limitations of the proposed method are fully discussed here.

• 융복합기술연구소 논문집
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• 제3권2호
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• pp.1-6
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• 2013

The ionic liquids are known to potential alternative solvents capable of replacing the commercial solvents in various processes including those in nuclear fuel cycle. As to the material, a number of studies have already reviewed the interesting results and addressed the spectroscopic as well as electrochemical behaviors of metal elements included in spent nuclear fuels. It has found that the important properties of metal ions in TBP dissolved ILs have led the development of alternative technologies to traditional solvent extraction processes. On the other hand, the electrochemical deposition of metal ions in ILs have been investigated for the application of the solvents to aqueous as well as to non-aqueous processes. In this work, a review on the application of ILs in nuclear fuel cycle is briefly presented to understand the notable researches on ILs focusing on aqueous processes.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2682-2695
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• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.